1. Field of the Invention
The present invention relates to a miniaturized nuclear utilizing improved pressure tube structural members. More particularly, the present invention relates to a new miniaturized nuclear reactor utilizing novel structural members that are used to support the loads and stresses of multiple nuclear reactor fuel channel pressure tubes in a confined area.
2. Description of the Prior Art
Nuclear power plants traditionally have been designed for achieving long term, safe, and reliable performance. To assure safety, the plants incorporate systems and procedures representing a studied anticipation of emergency conditions. design approaches will have considered theories or premises which may include, for example, design redundancies which are challenged by updated rules of performance as operating experience with nuclear power progresses. Thus, investigators in this power field continuously are called upon to develop improved analytic models of operation exhibiting improved bounding of operational factors and to further achieve higher levels of safety in view of changing rules of safety related performance. Because of the necessarily extensive time interval involved in developing or constructing a new nuclear power facility, for example such an effort may encompass ten years or more, and further in view of the numerous nuclear power facilities now in operation, these investigators typically are called upon to meet new rule criteria by modification of longexisting facilities. Retrofitting procedures can be quite extensive, calling for revised electrical power supplies, major valving replacements, and the like.
The nuclear industry has evolved a variety of reactor types. One such type finding substantial field use performs to produce steam for turbine drive within the reactor core itself and is referred to as a boiling water reactor (BWR). The reactor heated water of the BWR serves not only as working fluid, but also as a reaction moderator, and along with other parameters, its proper supply and application within the system necessarily has been the subject of safety requirements or rule generations by government regulatory agencies such as the Nuclear Regulatory Commission (NRC).
Typically, the general structure of a BWR nuclear system will include an upstanding reactor vessel which incorporates a lower reactor core structure beneath which are control rod drives. Above the core are, in order, a steam separator assembly and a steam dryer assembly leading to a steam outlet, above the reactor is a shield wall and outwardly of that a drywell. A pressure suppression chamber (wetwell), being torroidal in shape, is located below and encircling the drywell.
In more typical BWR installations, water coolant is heated in the reactor core to rise within the reactor vessel as a two-phase mixture of water and steam. This dual phase mixture then passes upwardly through the steam separator assembly and steam dryer structure to enter the steam line leading to a turbine. Following turbine drive, the steam is condensed to water and returned to the reactor by relatively large condensate and feedwater pumps of a feedwater system. The feedwater enters the downcomer region of the reactor, where it is mixed with the water returning from the steam separator and drying functions. The water in the downcomer region is circulated through the reactor core via the vertically oriented recirculation pumps which direct flow to the vertical jet pumps located between the core shroud and vessel wall (downcomer annulus). In typical fashion, two distinct recirculation loops with corresponding recirculation pumps are employed for this recirculation function.
In the event of some form of breakage or excursion generating malfunction, referred to as a "loss-of-coolant accident" (LOCA), designers anticipate that the relatively higher temperature-higher pressure water within the reactor will commence to be lost. A variety of safety systems and procedures may then be invoked both for containment and for thermal control of this LOCA. For the latter, thermal control, safety designs recognize that, while loss of the water moderator terminates the core reaction to eliminate a possibility of a nuclear incident, the momentum of generated heat or the residual energy within the reactor will remain of such magnitude as to require a cooling control to avoid for example, core melt down. In general, the amount of water within the containment system is more than adequate for this purpose, for example that contained in the suppression pool, or additionally, the condensate storage tank. To apply this water coolant for the safety purpose, a variety of safety related techniques or "emergency core cooling systems" (ECCS) have been developed to accommodate the LOCA. For example, core spray (CS) systems and low pressure coolant injection (LPCI) installations have been evolved in a variety of configurations.
The LPCI system incorporates, for example, four pumps which are activated by a safety system in the event of a coolant loss. Where the loss of coolant is of sufficient extent, and the vessel pressure remains high, for example in the event of a small pipe break then, an automatic safety system will function to depressurize the reactor vessel permitting the relatively lower pressure water supply pumps to operate to introduce water to the reactor. Because the recirculation system earlier described is ideally structured for this purpose, generally it is used by the LPCI system for water introduction under ECCS conditions.
Safety designs heretofore have recognized, however, that a recirculation loop may be broken under a LOCA condition. Thus, the pumping of water into that loop under such a LOCA condition may have no effectiveness. Accordingly, the LPCI systems have been equipped with a recirculation loop selection feature termed "loop selection logic" to avoid such conditions. This safety control detects the broken recirculation loop and initiates a procedure injecting water into the redundant, intact recirculation loop by actuating appropriate LPCI injection valves. Experience with such LPCI loop selection features have shown them to be complex and difficult to test and maintain. Under more current rule-based requirements, the design must accommodate for such occurrences as valve failure and the like. However, to function more effectively under current rules, procedures for retrofitting existing facilities to update them are elaborate and quite expensive, implementation involving such activities as recabling, pump reconnection activities and the like. Thus, an approach has been sought by investigators which offers operators the opportunity to eliminate the requirement for a loop selection logic regimen and associated costs therewith while improving the reliability of the LPCI system.
Numerous innovations for structural member for nuclear reactor pressure tubes have been provided in the prior an that are described as follows. Even though these innovations may be suitable for the specific individual purposes to which they address, they differ from the present invention as hereinafter contrasted.
In U.S. Pat. No. 3,584,903 titled ROLLED CHANNEL JOINTS by inventor James David Prichard, a strong and leak-free hub assembly for use with the pressure tubes of a nuclear reactor is disclosed in which the hub includes a hard insert having at least one groove formed in it, the hardness of the insert being greater than the hardness of the tubular element with which it is joined. Typically, the hub is formed of stainless steel, the insert is formed of surface hardened stainless steel and the tubular element is a zirconium-niobium alloy. The insert has a hardness greater than the hardness of the tubular element. The present invention differs from the above described patent due to the features of the joint assembly, the present invention utilizes a threaded joint connector to join a fuel channel pressure tube to the reactor system.
In U.S. Pat. No. 4,555,361 titled METHOD OF REDUCING THE VOLUME OF SOLID RADIOACTIVE WASTE by inventor Leo P. Buckley et at., combustible, solid radioactive waste, such as paper, plastics, rubber, cloth and wood are reduced in volume to ash residue using pyrohydrolysis, a method which combines pyrolysis of the waste in a vessel at temperatures in the range of 500.degree. to 700.degree. C. and gasification of residual carbon with superheated steam. Pressures of 1.0 to 3.5 Mpa are used with steam flows in the range 4 to 50 grams/second/cubic meter so that carbon containing components of the waste are removed as gaseous decomposition products in the form of carbon monoxide and hydrogen leaving an ash residue. The present invention differs from the above described patent due to the features of a method of reducing the volume of ash produced whereas the present invention describes utilizing glass and other impurities which when reacted with the fuel pellets form a less radioactive substance.
In U.S. Pat. No. 4,627,069 titled JOULE MELTER FOR THE PROCESSING OF RADIOACTIVE by inventor Keith B. Harvey et at., the joule melter has an outer cylindrical electrode which forms the outer wall of the melt containment, an inner cylindrical electrode which protrudes upward in the containment and forms the outlet for the melt, thus, also determining the depth of the melt. A non-conducting sealing material forms a base plug between the electrodes. A cylindrical electrically conductive baffle is located between the electrodes and includes an opening which allows the melt to flow from near the outer electrode where the melt material is first inserted into the melter, to the inner electrode which is the outlet. In addition to the inner and outer electrodes, the baffle may be connected to a power supply to modify the currents flowing at each of the electrodes. The present invention differs from the above described patent due to the features of melting the radioactive waste whereas the present invention describes utilizing glass and other impurities which when reacted with the fuel pellets form a less radioactive substance.
In U.S. Pat. No. 3,837,397 rifled TUBE BUNDLE ASSEMBLY by inventor Michael J. Pettigrew, a robe bundle assembly, for example, a heat exchanger tube bundle or a nuclear fuel element tube bundle, comprises a bundle of laterally spaced tubes, a frame around the outermost tubes, and a lattice of wire cables with their ends held against lateral displacement by the frame and the tubes in the lattice interstices. The cables are deflected round a portion of each tube to space the tubes from one another, and the cables are preferably tensioned against the frame for this purpose. The present invention differs from the above described patent due to the features of the cable matrix whereas the present invention describes utilizing a moderator comprising separate compartments within which the calandria tubes and fuel channel pressure tubes are contained. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein.
In U.S. Pat. No. 5,213,757 rifled METHOD FOR FIXING A SPRING PACKAGE TO A TOP NOZZLE IN A FUEL ASSEMBLY OF A NUCLEAR POWER REACTOR by inventor Lennart Ohman, a method of fixing a spring package to a top nozzle in a fuel assembly of a nuclear reactor wherein the fuel assembly comprises fuel rods, guide tubes and spacers arranged in a bundle between a top nozzle and a bottom nozzle wherein a T-shaped slot in milled out in a clamp which is welded to or forms an integral part of the top nozzle for receiving one end of the spring package, the end of the spring package is then inserted into the slot and the end is then fixed in the slot by means of a locking pin. The present invention differs from the above described patent due to the features of the bundle whereas the present invention describes utilizing a moderator comprising separate compartments within which the calandria tubes and fuel channel pressure tubes are contained. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein.
In U.S. Pat. No. 5,213,755 titled LOW PRESSURE COOLANT INJECTION MODIFICATION FOR BOILING WATER REACTORS by inventor David M. Kelly et al., a conventional low pressure coolant injection system for boiling water reactors is provided. With the modification, the cross tie conduits and associated valves remain open between two LPCI divisions. On the occasion of an LOCA, the LPCI pumps are activated and injection valves for each of the LPCI divisions are opened to commence coolant injection in the recirculation loops in simultaneous fashion. However, the rate of flow of water coolant within each injection system is controlled by a hydraulic resistance, which is selected to achieve reactor core cooling within requisite quantifies from one injection path. Thus, even though coolant water may flow through a rupture within one recirculation loops, sufficient water will be injected in the other loop to achieve core cooling. The present invention differs from the above described patent due to the features of the coolant system whereas the present invention describes utilizing a moderator comprising separate compartments having coolant systems flowing therein within which the calandria tubes and fuel channel pressure tubes are contained and are maximally cooled due to the novel features of the present invention. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein which achieve this cooling maximization.
In U.S. Pat. No. 4,788,033 titled CALANDRIA by inventor Luciano Veronesia calandria for use in conducting the hot coolant of a nuclear reactor transversely. The calandria includes an upper plate and a lower plate which support tubes. The plates and tubes are enclosed in a shell which extends above the upper plate and has a supporting flange. The lower plate has holes for transmitting coolant into the region between the plates. The shell has openings whose boundaries mate with the outlet nozzles of the reactor. The tubes are of stainless steel and are dimensioned so that they have mass, stiffness and strength such that they are not subject to failure by the transverse flow of the coolant even at a high velocity. The present invention differs from the above described patent due to the features of the coolant system whereas the present invention describes utilizing a moderator comprising separate compartments having coolant systems flowing therein within which the calandria tubes and fuel channel pressure tubes are contained and are maximally cooled due to the novel features of the present invention. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein which achieve this cooling maximization.
In U.S. Pat. No. 4,788,032 rifled REACTOR WITH FLOW GUIDANCE IN THE UPPER INTERNALS by inventor Jacques Baujat et al., a nuclear reactor has a pressure resistant vertical vessel with inlet and outer pipes situated at the same horizontal level. It also includes internals having a barrel supporting the core and defining with the vessel a down flow path for the coolant from the inlet pipes towards a space under the core and upper internals defining a flow path for the coolant leaving the core, above the latter, and flowing towards the outlet pipes. The upper internals include dividing walls defining circumferentially distributed volumes located at the common level of the pipes and each over a different angular sector. Some volumes belong to the initial part of the down going coolant path and the others force part at least of the coolant leaving the core to follow a path which is successively directed upwardly then curving towards the outlet pipes. The invention is particularly suitable for use in pressurized water reactors. The present invention differs from the above described patent due to the features of the coolant system whereas the present invention describes utilizing a moderator comprising separate compartments having coolant systems flowing therein within which the calandria tubes and fuel channel pressure tubes are contained and are maximally cooled due to the novel features of the present invention. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein which achieve this cooling maximization.
In U.S. Pat. No. 4,759,904 rifled PRESSURIZED WATER REACTOR HAVING IMPROVED CALANDRIA ASSEMBLY by inventor James E. Gillet et al., a calandria assembly is received within the pressure vessel of a nuclear reactor system, at an elevation corresponding to the level of the outlet nozzles of the vessel, and receives pressurized coolant traveling in an axial flow direction within the vessel and turns same to a radial direction for exit though the outlet nozzles. Hollow tubes mounted in parallel relationship at opposite ends to first and second plates of the calandria in conjunction with a cylindrical skin of cylindrical configuration joining the first and second plates of the calandria, present a redundant structure introducing the potential of thermal stresses, which are limited by selection of the pattern of flow holes in the lower plate and the provision of flexible annular weld joints of J-shaped configuration between the lower ends of the calandria tubes and the lower, second calandria plate. The present invention differs from the above described patent due to the features of the coolant system whereas the present invention describes utilizing a moderator comprising separate compartments having coolant systems flowing therein within which the calandria tubes and fuel channel pressure tubes are contained and are maximally cooled due to the novel features of the present invention. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein which achieve this cooling maximization.
In U.S. Pat. No. 4,284,475 rifled WEAR SLEEVE FOR CONTROL ROD GUIDE TUBE by inventor Andrew J. Anthon, a wear sleeve for a guide robe in a nuclear fuel assembly, and a method of installing the sleeve. The sleeve is an elongated metal cylinder having an upper portion adapted to be suspended from the upper end of the guide tube, and a lower portion adapted to be permanently deformed into interference fit with the walls of the guide tube whereby the sleeve may be secured against vertical movement. The method of installing the sleeve includes the steps of suspending the sleeve from the upper end of the guide tube, then expanding a selected lower surface of the sleeve until the sleeve is permanently deformed, whereby an interference fit between the sleeve and robe is formed. The present invention differs from the above described patent due to the features of the joint assembly, the present invention utilizes a threaded joint connector to join a fuel channel pressure tube to the reactor system.
Numerous innovations for structural member nuclear reactors utilizing nuclear reactor pressure tubes have been provided in the prior art that are adapted to be used. Even though these innovations may be suitable for the specific individual purposes to which they address, they would not be suitable for the purposes of the present invention as heretofore described.